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Sensitivity and Uncertainty Quantification of Neutronic Integral Data Using ENDF/B-VII.1 and JENDL-4.0 Evaluations

Mustapha Makhloul, Hamid Boukhal, Tarek El Bardouni, El Mahjoub Chakir, Mohamed Kaddour and Sanae Elouahdani

A chapter in Nuclear Power Plants - The Processes from the Cradle to the Grave from IntechOpen

Abstract: Many integral neutronic parameters such as the effective multiplication factors (keff) are based on neutron reactions with matter through cross sections. However, these cross sections present uncertainties, of origin multiple, which reduce the safety margin of nuclear installations. In order to minimize these risks, a sensitivity analysis is necessary to indicate the rate of change of a reactor performance parameter compared to variations in cross sections. Thus, several critical benchmarks were taken from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE), and their sensitivities and covariance matrix of the desired cross section were processed by MCNP6 and NJOY codes, respectively, in ENDF/B-VII.1 and JENDL-4.0 evaluations. The results obtained show that the 44 energy groups give the most varied sensitivity profiles than those given by others (15 and 33). In addition, we observed large uncertainties on the keff due to the H-1 and O-16 cross-sectional uncertainties (~200-1000 pcm) in ENDF/B -VII.1 and the U-235 cross section in JENDL-4.0; however, keff's uncertainties due to the cross-sectional uncertainties of the U-238 are very small.

Keywords: keff; sensitivity; covariance matrix; uncertainty; MCNP6.1; NJOY; multigroup cross section (search for similar items in EconPapers)
JEL-codes: Q40 (search for similar items in EconPapers)
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Persistent link: https://EconPapers.repec.org/RePEc:ito:pchaps:212955

DOI: 10.5772/intechopen.92779

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